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DOE OSTI2017

Storage, transportation and disposal system for used nuclear fuel assemblies

An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Storage, transportation and disposal system for used nuclear fuel assemblies

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Innovative Approach to SCC Inspection and Evaluation of Canister in Dry Storage

This final report is summarized four years of work of awarded project number DE-NE0008442\IRP-15-9318, title as: “Innovative Approach to SCC Inspection and Evaluation of Canister in Dry” supported by Nuclear Energy University Program\Integrated Research Program, NEUP\IRP, where last year is an extension of no cost project. This research work is focused on the possible degradation of dry storage canister holding used nuclear fuel (UNF), due to exposure of marine atmospheric environment. The canister usually designed from a welded austenitic stainless-steel of types 304 or 316, encased in a concrete shell known as an overpack. This canister designed and licensed for 20 years with an extension period need to be determined, in addition 20 years or more. These casks are located throughout the country in Independent Spent Fuel Storage Installations (IFSFIs) often on the site close to nuclear power plants, that are often located in coastal or lake-side regions where a marine atmospheric salt air environment exist. As a result of several recent decisions, such as not pursuing Yucca Mountain as a long-term geologic repository storage facility for UNF and the cancellation of the construction of a reprocessing facility in the 1980s due to proliferation risks, interim dry storage facilities for long-term storage are vital option for the immediate future of nuclear energy technology. One of the primary concerns with respect to the long-term performance of the storage casks is the potential for corrosion initiation due to deliquescence of salts deposited on the canister surface as aerosols; in regions of high residual weld stresses, this may lead to localized stress corrosion cracking (SCC). Dust and aerosols in the air being drawn though ventilation openings in the overpacks of passively ventilated dry canister storage systems may be deposited on the stainless-steel canister outer surfaces. Under these conditions, localized corrosion attack can occur. Chlorides ions deliquescence on the canister surface, a process by which the salts essentially pull moisture out of the air, creating a concentrated salt brine on the canister surface. Chloride-induced stress corrosion cracking (CISCC) of welded zones is of special concern, as it a well-documented mode of attack for austenitic stainless steels marine environments, and many independent spent fuel storage installations (ISFSIs) are located in coastal areas. Recent canister inspections have shown that chloride salts are present on the surface of in-service canisters at near-marine settings. However, canister surface inspections of satisfactory resolution to detect SCC have never been carried out in details, because access to the canister surfaces through vents in the overpacks is extremely limited, and high radiation fields make removal of the canisters from the overpacks undesirable. Although no significant corrosion damage has not been detected up to date, there is concern that this may lead to pitting corrosion, which could in turn develop into stress corrosion cracking, causing canister structural failure. The initiation and propagation of pitting corrosion has been widely studied in full immersion conditions, but limited studies have been done on marine atmospheric conditions, which is a focus of this work, these includes laboratory and field experimentally and propagation rate modeling of pitting and transition to crack, and crack propagation rate. It is known, however, that the rate of atmospheric corrosion can be affected by variations in relative humidity (RH), salt density, salt type, droplet area, temperature and stainless-steel microstructure and residual stress particularly in welded zones of canisters. The integrated research program has identified potential deterioration mechanisms for the stainless-steel canisters containing the UNF in dry storage that require detailed research and investigation and will have an impact on the performance of long-term interim storage under the normal and extreme environmental conditions experienced during the duration of this storage. The prediction and monitoring of canister corrosion processes while in storage can provide important data for the assessment of interim storage performance and the safety to the public. This report comprises of mainly the published results in different journals and covering a wide range spectrum from susceptible stainless-steel types 304\316 microstructural analysis through the implant test at atmospheric environment, residual stress analysis of mockup canister and simulation, inspection methodology includes nonlinear ultrasound and vibro-thermography and stochastic modeling and simulation of pits propagation rates and transition to cracks. Description of field specimens installed at ISFSI of Main Yankee is given in the following chapters and will be analyzed in the coming future and serves as an important data for a stochastic analysis as it was drawn from the publications and reports described in this report. The purpose of this report is to bring together the project efforts that includes improving in our understanding of structural aging performance at current dry storage casks fleet at ISFSIs in the context of remedial maintenance and inspection applications technologies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Next Generation System Analysis Model: Recently Added Features and Future Plans

To better enable informed decision-making regarding the back-end of the nuclear fuel cycle, the Integrated Waste Management System (IWMS) program within the U.S. Department of Energy, Office of Nuclear Energy (DOE-NE) has been sponsoring the development and application of system analysis tools capable of analyzing various system options for the management of spent nuclear fuel (SNF) and high-level radioactive waste (HLW).With these tools, IWMS architecture analyses are being conducted to support the future deployment of a comprehensive nuclear waste management system that considers all major back-end aspects of the nuclear fuel cycle (i.e., transportation, storage, and disposal). The Next Generation System Analysis Model (NGSAM) is an agent-based simulation software tool expressly designed to be capable of modeling features within various IWMS architectures. NGSAM imports data from Oak Ridge National Laboratory (ORNL)’s unified database (e.g., historic assembly information, thermal profiles for assembly heat, and at-reactor dry storage loadings) to ensure that each simulation initializes with a realistic representation of the state of commercial SNF in the U.S. Recent major enhancements implemented into NGSAM in the period since NGSAM was last presented at the WM2019 conference include: • Tracking of railroad escort car acquisition and buffer car acquisition • The addition of heavy haul truck (HHT) and barge routes for some sites, as well as support for user-defined inter-modal routes • Updates to the logic that checks the transportation cask thermal limit maps prior to package transport • An allocation method that predicts when reactor sites would pack assemblies from their spent fuel pools for dry storage, and prioritizes shipments directly from the pools of those reactor sites in the preceding periods (before the predicted loadings to dry storage), thus reducing the number of casks loaded into dry storage at reactor sites • The addition of “reactor site family” operational limits to restrict the number of loads of SNF taken from the pool and from dry storage at a given reactor site each year • Added support for multiple canister loading map options and packages with multiple compatible transportation overpacks • Updates to the handling of non-commercial SNF, including a new database that contains data to support the updates • Updates to allow analysis of hypothetical scenarios which include repackaging at reactor sites, e.g., for possible comparative analysis with other scenarios • Implementation of additional output reports, or modification of existing ones • The ability to generate and implement user edits via the NGSAM website • The ability to model loading SNF from pool storage at an interim storage facility (ISF) into dry storage at the ISF • The ability to model consolidating SNF from different existing storage containers at a DOE site into the same DOE standard canister • The ability to model transferring SNF casks from one transportation mode to another, e.g., from HHT to rail, referred to as transloading. These new features have improved NGSAM capabilities and users’ experience with the model. Preliminary NGSAM requirements for modeling advanced reactor fuels, reprocessing, treatment, and conditioning were considered, and this paper describes them at a high level. Other nuclear fuel cycle system analysis tools developed under sponsorship of DOE-NE, like the VISION code developed at Idaho National Laboratory, might be better suited for initial high-level analysis of those technologies and advanced fuel cycles. As technologies are developed and system concepts evolve, NGSAM could provide value by providing more detailed modeling of transport, storage, and disposal of spent fuel and wastes from advanced reactors and advanced fuel cycles at the fuel element and waste container level.

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DOE OSTI2018

Post-Irradiation Examinations of High Burnup PWR Fuel Rods - Initial Results

The High Burnup Spent Fuel Data Project (herein referred to as the “project”), sponsored by the US Department of Energy (DOE) Office of Nuclear Energy (NE), is focused on experimentally evaluating the effects of long-term storage and transportation on high burnup (HBU) (defined as >45 gigawatt days per metric ton uranium) pressurized water reactor (PWR) spent nuclear fuel (SNF) [1]. To support this project, in the fall of 2017, an instrumented bolted-lid cask (project cask) at the North Anna reactor site was loaded with intact, HBU PWR fuel assemblies (project assemblies) having four different kinds of cladding and will be reopened in approximately 10-years for inspection. To support pre- and post-storage SNF rod characterization, 25 “sister rods” that are similar to the rods being stored in the project cask were extracted from seven different fuel assemblies at the North Anna nuclear power plant for post-irradiation examination (PIE).

Bevard, Bruce Balkcom↗

DOE OSTI2018

Modeling the Drying of a Mock Used Fuel Assembly with COBRA-SFS Cycle-4 - 18597

The processing and storage of nuclear fuel is an ever increasingly pressing issue. Around 70,000 tons of used fuel are currently in either dry or wet storage on site at nuclear power plants or at specified spent fuel storage facilities. Additionally, about 2,000 tons are added to that sum each year. To promote retrievability, limit degradation in the canister, and reduce potential for hydrogen generation, spent fuel that is initially in wet storage must be dried before moving to dry cask storage. Therefore, this research seeks to use a mock used fuel storage cask to replicate the conditions in the industry during the drying process of fuel being moved from storage pools. To simulate this, a mock storage cask has been constructed to closely resemble those used in industry such that any areas that tend to retain water are replicated effectively. It is able to hold one mock used fuel assembly. The assembly used is an AREVA 10A 10 x 10 assembly with 78 rods filled with depleted uranium, 12 heater rods to simulate decay heat, a 3 x 3 BWR water channel, and one interchangeable rod position for tests such as a simulated failed fuel rod, BWR water rod, and PWR guide tube tests. The industry standard procedure is used in drying the assembly which involves stepping the system pressure down and holding for certain periods until reaching 3 torr and holding for 30 minutes. This is believed to sufficiently dry the assembly. The goal is to record data throughout these tests that can provide the industry with a greater understanding of how well the process works. This data will also be used to develop and validate a computer model of the drying process. Mechanistic models are currently under development. Reduced order models are being investigated for incorporation into current related engineering codes such as COBRA-SFS. COBRA-SFS is a member of the COBRA family of codes which is mainly used in the evaluation of thermal hydraulic systems containing spent nuclear fuel. However, it does not currently have the capability to account for processes such as evaporation encountered during drying operations. Therefore, those physical models are to be added to the source code. At present a model of the experiment has been created using the COBRA-SFS Cycle-4 code. A number of steady state models along with transient models have been developed and presented. Each of these yield valuable results for comparison with the experiments conducted. In addition to comparison with experimental results, the data collected from COBRA-SFS is compared with that which has been generated from other computer models which utilize computational fluid dynamics. (authors)

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DOE OSTI2016

GNS Castor V/21 Headspace Gas Sampling 2014

Prior to performing an internal visual inspection, samples of the headspace gas of the GNS Castor V/21 cask were taken on June 12, 2014. These samples were taken in support of the CREIPI/Japanese nuclear industry effort to validate fuel integrity without visual inspection by measuring the 85 Kr content of the cask headspace

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DOE OSTI2016

Signatures of Extended Storage of Used Nuclear Fuel Comprehensive Final Report

This report serves as a comprehensive overview of the Extended Storage of Used Nuclear Fuel work performed for the Material Protection, Accounting and Control Technologies campaign under the Department of Energy Office of Nuclear Energy. This paper describes a signature based on the source and fissile material distribution found within a population of used fuel assemblies combined with the neutron absorbers found within cask design that is unique to a specific cask with its specific arrangement of fuel. The paper describes all the steps used in producing and analyzing this signature from the beginning to the project end.

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DOE OSTI2016

Partial defect verification of spent fuel assemblies by PDET: Principle and field testing in Interim Spent fuel Storage Facility (CLAB) in Sweden

Here, the need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called "difficult-to-access" areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into "difficult-to-access" areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reported the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17×17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly bunrup levels.

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DOE OSTI2015

The radiation characteristics of the transport packages with vitrified high-level waste

The calculation method of neutron yield in the (α, n) reaction for a homogeneous material of arbitrary composition is represented. It is shown that the use of the ORIGEN 2 code excluding the real elemental composition of vitrified high-level waste leads to significant underestimation of the neutron yield in the (α, n) reaction. For vitrified high-level waste and spent nuclear fuel from VVER, the neutron fluxes are analyzed. The thickness of the protective materials for a transfer cask and a shipping cask with vitrified highlevel waste are estimated.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2014

Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

This report presents a preliminary evaluation of removing used nuclear fuel (UNF) from 12 shutdown nuclear power plant sites. At these shutdown sites the nuclear power reactors have been permanently shut down and the sites have been decommissioned or are undergoing decommissioning. The shutdown sites are Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, and San Onofre. The evaluation was divided into four components: characterization of the UNF and greater-than-Class C low-level radioactive waste (GTCC waste) inventory; a description of the on-site infrastructure and conditions relevant to transportation of UNF and GTCC waste; an evaluation of the near-site transportation infrastructure and experience relevant to shipping transportation casks containing UNF and GTCC waste, including identification of gaps in information; and, an evaluation of the actions necessary to prepare for and remove UNF and GTCC waste. The primary sources for the inventory of UNF and GTCC waste are the U.S. Department of Energy (DOE) RW-859 used nuclear fuel inventory database, industry sources such as StoreFUEL and SpentFUEL, and government sources such as the U.S. Nuclear Regulatory Commission. The primary sources for information on the conditions of site and near-site transportation infrastructure and experience included observations and information collected during visits to the Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion sites; information provided by managers at the shutdown sites; Facility Interface Data Sheets compiled for DOE in 2005; Services Planning Documents prepared for DOE in 1993 and 1994; industry publications such as Radwaste Solutions; and Google Earth. State and Regional Group representatives, a Tribal representative, and a Federal Railroad Administration representative participated in six of the shutdown site visits. Every site was found to have at least one off-site transportation mode option for removing its UNF and GTCC waste; some have multiple options. Experience removing large components during reactor decommissioning provided an important source of information used to identify the transportation mode options for the sites. Especially important in conducting the evaluation were site visits, through which information was obtained that would not have been available otherwise. Extensive photographs taken during the site visits proved to be particularly useful in documenting the current conditions at or near the sites. Additional conclusions from this evaluation include: The 12 shutdown sites use designs from 4 different suppliers involving 9 different (horizontal and vertical) dry storage systems that would require the use of 8 different transportation cask designs to remove the UNF and GTCC waste from the shutdown sites; Although there are common aspects, each site has some unique features and/or conditions; Although some regulatory actions will be required, all UNF at the initial 9 shutdown sites (Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion) is in licensed systems that can be transported, including a small amount of high-burnup fuel; Each site indicated that 2-3 years of advance time would be required for its preparations before shipments could begin; Most sites have more than one transportation option, e.g., rail, barge, or heavy haul truck, as well as constraints and preferences. It is expected that additional site visits will be conducted to add to the information presented in the evaluation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Spent Nuclear Fuel Transportation: Rail Conveyance Transmissibility

Equipos Nucleares, S.A. (ENSA) and the US Department of Energy (DOE) are preparing a full scale Spent Nuclear Fuel (SNF) transportation test. This transportation testing will include road, rail, coastal, and trans-Atlantic shipments. The test campaign will use a full scale commercial transportation package and cradle. The package will be loaded with two instrumented fuel assemblies to measure strains at cladding locations and accelerations on the fuel assemblies, and thirty “dummy” assemblies. This testing is designed to closely match an actual SNF shipment. Accelerometers will also be used at various locations throughout the full conveyance system (i.e. rail car/truck, cask, transport cradle, and basket) to study the transmission of loads through the system and to provide validation for numerical models. Previous testing and modeling work has shown how the structural transmissibility of the transport system can affect the magnitude of these loads, and the importance of modeling all aspects of the transport system (i.e. rail car/truck, transport cradle, cask, basket, and fuel) (Ref 1, 2). This paper describes preliminary models that were constructed to estimate load transmission during rail transport, from the bottom of the cradle to an individual fuel rod within the package. The modeling studies in this paper evaluate the system response to postulated shock pulses and random vibration loads. These models describe the transmissibility of the conveyance, and demonstrate how loads can be amplified or attenuated as they are transmitted through the structure. This is done by coupling the rail vehicle dynamics code NUCARS, to the general finite element modeling code ANSYS, and the explicit dynamics code LS-DYNA. Models such as the ones presented herein will be used during the test campaign to help analyze and evaluate the test data as it is collected.

Cladding, Spent Fuel and Radiactive Waste, Used Fu↗

DOE OSTI2017

A Cable Car Transfer System for the Cigeo Access Ramps - An Original Approach - 17035

Andra is currently implementing the detailed engineering phase of the step-wise design of Cigeo (the French HLW and IL-LLW Deep Geological Repository). Various technical issues are at stake, including the definition and development of a key component: a funicular ('cable-car') system dedicated to the transfer (down the access ramp and leading to the underground infrastructures) of the shielding casks housing the waste disposal containers (packages). The present paper starts with a brief description of the Cigeo architecture (i.e. the surface utilities, the service and nuclear ramps, the various shafts and finally the underground facilities). A focus is then made on the 'nuclear ramp' dedicated to the transfer of the waste containers lodged inside the casks (those latter weighing some 100 to 130 metric tons each) and its main geometrical characteristics (slope, length, diameter, structure). Then a discussion follows to detail the rationale having led to the choice of a somehow 'sophisticated' funicular transfer system, at the expense of what could be called 'more conventional' and more straightforward transport solutions (like a transfer system by trucks or by rack and pinion trains). A systematic comparison is finally made between the various solutions, explaining and rating the pros and cons of each concept, via a multi-criteria analysis. This technical choice (and the related justification approach) was submitted to an Experts review at the end of the Preliminary Design phase of Cigeo and comforted by their strong support before the design freeze of this solution was integrated in Cigeo. The following part of the paper is then focused on a description of the main technical components and the mechanical performance of the cable-car system, its current state of design and the set of qualification tests (of key safety components) to be implemented before its effective construction. This article concludes with a timeline of the qualification tests and developments to come. The positive outcomes of this future technical test campaign will also pave the way for the future instruction of the Cigeo licensing file, scheduled as of mid-2018. (authors)

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DOE OSTI2015

Solutions for management of defective fuel assemblies - 15318

Defective fuel management is a major challenge for nuclear operators when the time comes for finding a long-term solution for their used (or spent) fuel assemblies. This question of defective fuel management can be raised in several cases: - For optimizing the remaining operating costs of shutdown reactors, the spent fuel pools must be emptied as soon as possible; a solution needs therefore to be found for evacuating all the spent fuel assemblies, defective fuel assemblies and in particular defective fuel rods included; - Some utilities are today facing a saturation of their reactor spent fuel pool; evacuating the defective fuel assemblies and extracted defective fuel rods allows optimizing the space in the pool. Whereas the solutions for managing intact used fuel assemblies are well-known, a specific technical solution must be implemented for managing the defective spent fuel assemblies as well. The purpose of this paper is to describe the existing solutions for the management of all types of defective fuel assemblies. Interim Dry Storage Solution Interim dry storage of defective spent fuel assemblies is being implemented for decades in the USA. Specific operations are performed for preparing the defective fuel assemblies for interim dry storage and they are then stored with intact used fuel assemblies in the dry storage cask. The advantage of such a solution is its short-term cost effectiveness. The paper presents the existing dry interim storage solutions for defective used fuel rods, the way they are implemented, the associated operational experience. Long-term Dry Storage Solution A long-term dry storage solution for defective fuel assemblies is being developed, compliant with direct disposal requirements. The solution is based on the encapsulation of each defective fuel rod into a fuel rod capsule and an associated specific dry storage licensing approach: this is the best available technology regarding long term storage safety requirements. The paper presents this unique capsule technology, the transport and storage casks that can be used for transportation of defective fuel rods to the storage facility, and the associated transport licensing approach currently used in Europe. Treatment Solution. The most definitive way to get rid of defective fuel assemblies is their treatment (or reprocessing) in a treatment plant. There is then no need to assess the behaviour of these specific used fuel assemblies as for an interim storage facility or a final repository. The advantage of such a solution is its sustainable approach. The paper presents the existing transport and treatment solutions, the way they are implemented, the associated operational experience. (authors)

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DOE OSTI2015

Development of the computer code for multi-calculations of source term of spent fuel - 15351

Many PWR spent fuels have been transferred between the neighboring nuclear power plants at Kori, Hanbit, Hanwul site, and from nuclear power plant to the KAERI(Korea Atomic Energy Research Institute) for the research purpose in Korea. In case of the PHWR type spent fuels, these have been transferred from the spent fuel pool to an on-site dry storage facility. Total radioactivity of spent fuels has to be calculated before transport in advance by the regulation requirement. In case of PWR, KN-12 transport cask can transfer 12 spent fuel assemblies at one time and in case of PHWR, Hi-STAR 63 transport cask can transfer 120 spent fuel assemblies at one time. Therefore, total radioactivity of all transported spent fuel assemblies that are transported generally are calculated by using the ORIGEN-ARP to meet this regulatory requirement. Basically, ORIGEN-ARP calculates the one fuel assembly at one time. SCALE code including this code provides the batch mode to perform the multi-calculations of the source term. In this study, in order to use the batch mode of ORIGEN-ARP, ARP-auto program was developed to generate input files and batch file of ORIGEN-ARP automatically and to make the summary file extracting the result from output files of ORIGEN-ARP. This code enhances the efficiency at the field. The working time to complete the calculation of source terms of 120 spent fuel assemblies take only 20 minutes. In order to confirm the credibility of this code, total radioactivity of 12 spent fuel assemblies of 17x17 Westinghouse type were calculated and compared to the previous result. The initial enrichments of these spent fuels are 3.46 ∼ 3.50 w/o and burnups are 29 ∼ 43 GWD/MTU. As the results, the relative errors are about 0.9 ∼ 3.0%. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Higher harmonic ultrasonic guided waves for structural integrity assessment of dry storage canisters - 15381

Dry storage of used fuel from nuclear power plants has gone from an interim storage solution to a longer-term solution, which necessitates monitoring the integrity of the stainless steel canister that confines the used fuel. Since the storage facilities are located around the country in a variety of different environments there is a wide range of degradation mechanisms that could occur over time in the cylindrical canister. One potential degradation mode that has created significant concern is stress corrosion cracking. We propose using higher harmonic ultrasonic guided waves as a tool to nondestructively inspect used fuel canisters. Higher harmonics are generated when a wave propagates through a nonlinear elastic media. These higher harmonics have been shown to be quite sensitive to microstructural features such as dislocation substructures, persistent slip bands, precipitates, etc. The use of ultrasonic guided waves to inspect the canister is attractive due to the size of the canister and the ability of guided waves to inspect a reasonably large domain relative to traditional bulk wave ultrasonics. We will highlight our recent progress on theory, numerical simulations, and laboratory experiments to develop higher harmonic guided waves as a tool to detect microstructural changes that precede macro-scale damage. Analyzed structures include plates and hollow cylinders, and we focus on localized damage initiation, as is the case for stress corrosion cracking. Recent experiments focus on using magnetostrictive transducers in activating the fundamental shear horizontal mode (SH0) and the third harmonic that is generated due to the material nonlinearity. The magnetostrictive transducer comprises a thin iron-cobalt foil, a meandering electric coil, and a permanent magnet. These transducers are inexpensive, easy to use, and are very efficient at generating shear waves. In applications where the wavelength is larger than the plate thickness, the waves are guided by the traction free boundary conditions and propagate in the plane of the plate. For this canister application, the shear horizontal waves energize the entire thickness of the canister and propagate along the shell in the direction dictated by the meander coil. Experiments indicate that magnetostrictive transducers can generate waves that propagate in plate thicknesses of 1 mm to at least 16 mm. The SH0 mode is nondispersive and generates a third harmonic of the same mode that is shown to be sensitive to localized plastic strain and unseen fatigue damage. It has a strong potential to detect the very early stages of stress corrosion cracking. Due to the geometric configuration of dry storage casks, the transducers will need to be delivered to the inspection points by a robotic system that navigates through the cask ventilation system. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Laying the groundwork for a large-scale spent nuclear fuel transportation system - 15408

The US Department of Energy Office of Nuclear Energy established the Nuclear Fuels Storage and Transportation Planning Project to lay the groundwork for implementing an interim storage facility, including associated transportation activities. Efforts include the development of a system for the large-scale transport of spent nuclear fuel that will be necessary in an integrated waste management system. Progress is being made on long lead time, destination-independent aspects of the transportation infrastructure. The large-scale transportation system for spent nuclear fuel is divided into three elements: institutional, operational, and hardware. The institutional element refers to the various forms of stakeholder interaction that must occur for this type of transportation system to be successful. It includes activities like development of a national transportation plan, work on policy development for Section 180(c) of the Nuclear Waste Policy Act, and identification of a preliminary suite of national transportation routes that reflect the interests of a broad cross section of stakeholders while meeting regulatory requirements. The operational element refers to the activities that must be undertaken to run a large-scale transportation system. This element is currently focused on development of a new SNF transportation routing analysis tool, study of the infrastructure near SNF storage sites that may be de-inventoried first, and development of tools for modeling transportation activities. The hardware element refers to the casks, railcars, and other items necessary to operate the system. This element currently focuses on development of railcars compliant with Association of American Railroads Standard S-2043, as well as studies related to the use of rail casks and their ancillary equipment. The Nuclear Fuels Storage and Transportation Planning Project is making significant progress in all three of these areas along the path forward to a fully operational transportation system. (authors)

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DOE OSTI2015

Achievements and process improvements in processing transuranic waste at the TRU Waste Processing Center - 15488

Wastren Advantage, Inc., (WAI) is the DOE Prime contractor for the TRU Waste Processing Center (TWPC) located in Oak Ridge, TN. This paper highlights the TWPCs specific achievements and process improvements associated with both Contact (CH) and Remote Handled (RH), Transuranic (TRU) waste receipt processing and repackaging for disposal. The five key improvement areas and achievements include: - The TWPC installed, tested and operated successfully the Drum Venting System (DVS) to vent over-packed CH-TRU waste drums in several configurations including 55-gallon drums over-packed in 79 or 85-gallon drums. - A new Remote Drum Opener (RDO) system was designed, tested and placed in service to vent drums discovered in overpack boxes in the Box Breakdown Area and to vent overpack configurations that could not be handled in the standard DVS (e.g., 110 gallon overpacks and multiple drum configurations). - The Cask Processing Enclosure (CPE) approach to repackaging RH-TRU waste contained in concrete casks was developed and implemented - Hot Cell outage to incorporate planned equipment modifications and improvements to maintain processing rates. - To improve certification and segregation of CH-TRU from drums that meet Low Level Waste criteria for disposal at Nevada the TWPC added additional equipment/facilities for Non-Destructive Assay (NDA) and Real Time Radiography (RTR). (authors)

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